I N T E R N A T I O N A L A T O M I C E N E R G Y A G E N C Y
T R A I N I N G C O U R S E S E R I E S 23
Boiling Water Reactor Simulator
Workshop Material
Second Edition
TRAINING COURSE SERIES No. 23
Boiling Water Reactor Simulator
Workshop Material Second Edition
INTERNATIONAL ATOMIC ENERGY AGENCY, 2005
The originating Section of this publication in the IAEA was:
Nuclear Power Technology Development Section International Atomic Energy Agency
Wagramer Strasse 5 P.O. Box 100 A-1400 Vienna, Austria
BOILING WATER REACTOR SIMULATOR IAEA, VIENNA, 2005
IAEA-TCS-23/02 ISSN 1018–5518
© IAEA 2005
Printed by the IAEA in Austria September 2005
FOREWORD
The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and workshop material and sponsors workshops. The workshops are in two parts: techniques and tools for reactor simulator development, and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No. 12, Reactor Simulator Development: Workshop Material (2001). Course material for workshops using a WWER-1000 simulator from the Moscow Engineering and Physics Institute, Russian Federation is presented in the IAEA publication: Training Course Series No. 21, 2nd edition, WWER-1000 Reactor Simulator: Workshop Material (2005). Course material for workshops using a pressurized water reactor (PWR) simulator developed by Cassiopeia Technologies Incorporated, Canada, is presented in the IAEA publication: Training Course Series No. 22, 2nd edition, Pressurized Water Reactor Simulator: Workshop Material (2005).
This report consists of course material for workshops using a boiling water reactor (BWR) simulator. W.K. Lam, of Cassiopeia Technologies Incorporated, developed the simulator and prepared this report for the IAEA. The IAEA officer responsible for this publication was R.B Lyon from the Division of Nuclear Power.
EDITORIAL NOTE
The use of particular designations of countries or territories does not imply any judgement by the publisher, the IAEA, as to the legal status of such countries or territories, of their authorities and institutions or of the delimitation of their boundaries.
The mention of names of specific companies or products (whether or not indicated as registered) does not imply any intention to infringe proprietary rights, nor should it be construed as an endorsement or recommendation on the part of the IAEA.
CONTENTS
1. INTRODUCTION ...1
1.1 Purpose ...1
1.2 Historical background ...1
1.3 Prominent characteristics of the BWR ...3
2. 1300 MW(E) BOILING WATER REACTOR NPP SIMULATOR ...5
2.1 Simulator startup ...7
2.2 Simulator initialization...7
2.3 List of BWR simulator display screens ...8
2.4 Generic BWR simulator display common features ...8
3. SIMULATOR DISPLAY SCREENS ...10
3.1 BWR plant overview screen...10
3.2 BWR control loops screen...12
3.3 BWR power/flow map & controls...14
3.4 BWR reactivity & controls screen...18
3.5 BWR scram parameters screen...20
3.6 BWR turbine generator screen ...21
3.7 BWR feedwater and extraction steam screen...22
4. SIMULATOR EXERCISES ...24
4.1 Introductory exercises...24
4.1.1 Power maneuver: 10% power reduction and return to full power....24
4.1.2 Reduction to 0% full power and back to 100% full power ...24
4.1.3 Turbine trip and recovery ...25
4.1.4 Reactor scram and recovery ...25
4.2 Malfunction exercises...26
4.2.1 Loss of feedwater - both FW pumps trip...26
4.2.2 Increasing core flow due to flow control failure ...28
4.2.3 Decreasing core flow due to flow control failure ...29
4.2.4 Decreasing steam flow from dome due to pressure control failure..30
4.2.5 Increasing steam flow from dome due to pressure control failure ...30
4.2.6 Turbine throttle PT fails low ...32
4.2.7 Safety relief valve (SRV) on one main steam line fails open...33
4.2.8 Feedwater level control valve fails open ...34
4.2.9 Turbine trip with bypass valve failed closed...35
4.2.10 Inadvertent withdrawal of one bank of rods...36
4.2.11 Inadvertent insertion of one bank of rods...37
4.2.12 Inadvertent reactor isolation ...38
4.2.13 Loss of feedwater heating...39
4.2.14 Power loss to three reactor internal pumps (RIPs) ...41
4.2.15 Steam line break inside drywell ...42
4.2.16 Feedwater line break inside drywell...43
4.2.17 Reactor vessel bottom break - 3000 kg/sec LOCA ...45
4.2.18 Load rejection...47
5. STEADY STATE MODEL ...48
5.1 Purpose ...48
5.2 Boiling water reactor mass and energy balance...48
5.3 Boiling water reactor spreadsheet model...51
5.3.1 Procedures for spreadsheet model ...53
5.3.2 Steady state model solutions ...55
6. DYNAMIC MODEL DESCRIPTION ...56
6.1 Reactor model...57
6.2 Fuel heat transfer ...59
6.3 Decay heat model ...60
6.4 Coolant heat transfer ...61
6.5 Core hydraulics and heat transfer ...63
6.5.1 Incompressible flow for non-boiling region...63
6.5.2 Compressible flow for boiling region...65
6.5.3 Boiling boundary...66
6.5.4 Summary of multi-nodal approach for simulating core hydraulics and heat transfer ...66
6.6 Saturated enthalpy, saturated liquid density...67
6.7 Core exit enthalpy, core quality, void fraction ...68
6.8 Dome mass balance and energy balance ...68
6.9 Saturated steam density and dome pressure ...69
6.10 Driving pressure in boiling core ...70
6.11 Recirculation flow & pressure losses ...71
6.12 Coolant recirculation pumps ...72
6.13 Feedwater flow ...72
6.14 Main steam system ...73
6.15 Control and protection systems ...74
6.15.1 Control rods control system...75
6.15.2 Recirculation flow control...75
6.15.3 Reactor pressure control system ...76
6.15.4 Reactor water level control system...77
6.15.5 Turbine power control system ...78
6.15.6 Turbine steam bypass control system...78
6.15.7 Protection system...78
6.15.8 Automatic responses to design basis events accidents ...79
APPENDIX: BWR technical data...81
1. INTRODUCTION 1.1 PURPOSE
The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in reactor operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the supply or development of simulation programs and educational materials, sponsors workshops, and distributes documentation and computer programs.
This publication consists of course material for workshops using the boiling water reactor (BWR) simulator. Participants in the workshops are provided with instruction and practice in using the simulator, thus gaining insight and understanding of the design and operational characteristics of BWR nuclear power plant systems in normal and accident situations.
This manual is written with the assumption that the readers already have some knowledge of the boiling water reactor. Therefore no attempt has been made to provide detailed descriptions of each individual BWR subsystem. Those descriptions are commonly found in nuclear engineering textbooks, BWR nuclear power plant (NPP) design manuals, and IAEA technical publications. However, details are provided where necessary to describe the functionality and the interactive features of the individual simulator screens, which relate to the specific BWR subsystems.
The manual covers basic NPP operations, such as plant load maneuvering, trips and recovery e.g. turbine trip and reactor scram. In addition, it covers plant responses to malfunction events.
Some malfunction events lead to reactor scram or turbine trip. Other serious malfunction events (e.g. LOCA) lead to actuation of the core cooling safety system.
It should be mentioned that the equipment and processes modeled in the simulator represent realistic BWR characteristics. However, for the purpose of the educational simulator, there are necessary simplifications and assumptions made in the models, which may not reflect any specific vendor’s BWR design or performance.
Most importantly, the responses manifested by the simulator, under accident situations, should not be used for safety analysis purposes, despite the fact that they are realistic for the purpose of education. As such, it is appropriate to consider that those simulator model responses perhaps only provide first order estimates of the plant transients under accident scenarios.
1.2 HISTORICAL BACKGROUND
Boiling water reactor plants were designed in the 1950s and put into operation starting from the early 1960s. Many BWR plants have been constructed and operated safely worldwide.
They constitute a significant electricity source from nuclear fission.
Although changes and improvements have been made in BWR designs throughout their history of operation, the basic concept is essentially unchanged since the first BWR design proposed by General Electric.
The basic feature of all BWR plants is the presence of a reactor pressure vessel (RPV) in which feedwater enters the vessel in subcooled conditions and saturated steam exits the vessel. The subcooled feedwater is heated by nuclear fission heat from the fuel bundles, as it travels up the various coolant channels in the reactor core. As boiling of the reactor coolant occurs at the upper region of the core coolant channels, a water-steam mixture exits the reactor core (into the upper plenum) at saturation temperature. Nominal core operating pressure is typically 7.0 MPa, which is nearly the same for all BWR designs. The water-steam mixture is then separated and dried in the upper plenum, with the saturated steam flowing directly to the turbine. The saturated liquid, separated from the water-steam mixture, is then recirculated back to the annular downcomer of the RPVwhere the subcooled feedwater mixes with the saturated liquid in the lower plenum region of the reactor vessel, before entering the coolant channels.
Originating from this basic BWR core design are the following reactor types that have been built and operated. Examples of existing plants are shown in parentheses:
• Natural circulation direct cycle BWR (Dodewaard)
• Forced Circulation Duel Cycle (Dresden 1, Garigliano)
• Forced Circulation Direct Cycle - external pumps (Ringhals, Oyster Creek)
• Forced Circulation Direct Cycle - jet pumps (Leibstadt, Dresden 2)
• Forced Circulation Direct Cycle - internal pumps (Kashiwazaki-Kariwa, Oskarshamn) The evolution of the BWR design was started with a relatively complex dual cycle design (Dresden 1), involving an intermediate steam generator. Then the design evolved to a direct cycle using external pumps (e.g. Oyster Creek), to a jet pump reactor (e.g. Dresden 2).
Nowadays, an example of a new BWR is the advanced boiling water reactor (ABWR) using internal pumps (e.g. Kashiwazaki-Kariwa), jointly designed by General Electric, Hitachi and Toshiba.
Historically, the dual cycle plant was designed and constructed in the early stages of BWR development, with the plant in the reactor-follow mode, i.e. the reactor power follows the turbine power. An intermediate steam generator was introduced to boil the feedwater utilizing the saturated liquid extracted from the primary circuit. The steam produced in the steam generator flowed to the turbine through a secondary steam line and was used to control the plant in reactor-follow mode. All the plants of this type are currently shutdown.
In order to increase the core cooling capability, pumps were introduced in the recirculation loop. The simplest configuration was with external pumps suctioning the fluid from the downcomer region and injecting it at higher pressure into the lower plenum of the core.
Further BWR development involved the recirculation loop configuration with jet pumps. The introduction of jet pumps satisfied two additional design objectives: (1) only a portion of the core coolant was recirculated externally to the vessel, and (2) no large pipe was connected to the bottom of the vessel, thus making core flooding easier in the unlikely event of a large pipe break in vessel-connected piping.
The configuration with internal pumps in the recirculation loop eliminates the piping and flows external to the vessel. Such a configuration is currently in the advanced BWR design.
A large variety of fuel designs is currently available. While the fuel box dimension generally remained unchanged, the number of fuel rods and rods lattice changed from 7 × 7, to 8 × 8, up
to 10 × 10. The control rods are inserted through the vessel from the bottom. As a consequence of larger moderation at the channel bottom, the axial power shape is typically bottom-skewed.
Finally, almost all BWR plants are equipped with a pressure suppression containment including a large pool of ambient temperature liquid (~ 5000 m3) where the steam-liquid mixture lost from a LOCA (loss of coolant accident) can be condensed. The containment also serves as a protective shield and prevents release of radioactive contamination to the outside of the reactor building, in the unlikely event of a serious accident.
1.3 PROMINENT CHARACTERISTICS OF THE BWR
A typical BWR is characterized by several prominent differences from other light water reactors (LWRs) such as the PWR:
(1) The core normal operating conditions result in two phases: a subcooled liquid phase in the non-boiling region; and a saturated steam-water mixture in the boiling region.
(2) Steam generation occurs in a direct cycle with steam separators and dryers inside the reactor pressure vessel. A separate steam generator is not required. Typical operating temperature is 288°C; steam pressure ~ 7 MPa.
(3) The reactor (steam dome) pressure is controlled by turbine inlet valves and turbine bypass valves.
(4) The BWR core consists of a number of fuel bundles (assemblies), each with a casing called a fuel channel. Each fuel bundle (assembly) contains a number of fuel rods arranged in a N × N square lattice, with slightly enriched Uranium fuel ~ 2% to 5% U- 235 by weight.
(5) The control rods are of cruciform shape and enter the core from the bottom. Each control rod moves between 4 fuel assemblies.
(6) The reactor power control consists of control rods and recirculation flow control.
Control rods are used to achieve the desired power level by adjustment of their positions in the core at a rate equivalent to a power change rate of 2% full power per sec. The recirculation flow control also controls reactor power by causing the density of the water used as moderator to change. The flow rate is adjusted by a variable speed pump (such as the internal pumps of the ABWR) at a rate equivalent to a power change rate of 30%
full power per minute.
(7) “Dried” steam from the reactor pressure vessel (RPV) enters the turbine plant through four steam lines connected to nozzles equipped with flow limiters. In the unlikely event of a steam line break anywhere downstream of the nozzle, the flow limiters limit the steam blowdown rate from the RPV to less than 200% rated steam flow rate at 7.07 MPa.
(8) There are safety relief valves (16 of them) connected to the four steam lines to prevent RPV overpressure, with a blow down pipe connected to the suppression pool.
(9) In the steam lines, isolation valves are provided inside and outside of the containment wall to isolate the RPV, if necessary.
(10) Saturated steam from the RPV main steam lines is then admitted to the turbine HP cylinder via the governor valves. After the HP section, steam passes through the moisture separator reheater (MSR) to the LP turbine cylinders.
(11) A special steam bypass line, prior to the turbine governor valves, enables dumping the full nominal steam flow directly to the condenser in the event of plant upset such as a turbine trip, in order to avoid severe pressure surges and corresponding power peaks in the reactor.
(12) Typical balance of plant (BOP) systems for the BWR consists of the condenser, condensate pumps, deaerator, feedwater heaters, reactor feed pumps (RFP) and reactor level control valves.
(13) The containment is a cylindrical prestressed concrete structure with an embedded steel liner. It encloses the reactor, reactor coolant pressure boundary and important ancillary systems. The containment has a pressure-suppression type pool with drywell and wetwell.
A typical BWR design with the above described features is shown in Figure 1. Descriptions of a number of BWR designs can be found in IAEA-TECDOC-1391 Status of Advanced Light Water Cooled Reactor Designs (2004).
FIG. 1. A typical 1300 MW(e) boiling water reactor NPP.
2. 1300 MW(e) BOILING WATER REACTOR NPP SIMULATOR
The purpose of the 1300 MW(e) boiling water reactor NPP simulator is educational — to provide a teaching tool for university professors and engineers involved in teaching topics in nuclear energy. As well, nuclear engineers, scientists and teachers in the nuclear industry may find this simulator useful in broadening their understanding of BWR NPP dynamics and transients.
The simulator can be executed on a personal computer (PC), to operate essentially in real time, and have a dynamic response with sufficient fidelity to provide BWR plant responses during normal operations and accident situations. It also has a user-machine interface that mimics the actual control panel instrumentation, including the plant display system, and more importantly, allows user interaction with the simulator during the operation of the simulated BWR plant.
The minimum hardware configuration for the simulator consists of a Pentium PC or equivalent (minimum 166 MHz CPU speed), minimum of 64 Mbytes RAM with 256 external Cache, at least 500 Mbytes enhanced IDE hard drive, 2 Mbytes VRAM, hi-resolution video card (capable of 1024 x 768 resolution), 15 inch or larger high resolution SVGA colour monitor, keyboard and mouse. The operating system can be Windows 95, Windows NT, Windows 2000, or Windows XP.
The requirement of having a single PC to execute the models and display the main plant parameters in real time on a high-resolution monitor implies that the models have to be as simple as possible, while having realistic dynamic response. The emphasis in developing the simulation models was on giving the desired level of realism to the user. That means being able to display all plant parameters that are critical to operating the unit, including the ones that characterize the main process, control and protective systems. The current simulator configuration is able to respond to the operating conditions normally encountered in power plant operations, as well as to numerous malfunctions, as summarized in Table I.
The simulation development used a modular modeling approach: basic models for each type of device and process are represented as algorithms and developed in FORTRAN. These basic models are a combination of first order differential equations, logical and algebraic relations.
The appropriate parameters and input-output relationships are assigned to each model as demanded by a particular system application.
The interaction between the user and the simulator is via a combination of monitor displays, mouse and keyboard. Parameter monitoring and plant operator controls, ,are represented in a virtually identical manner on the simulator. Control panel instruments and control devices, such as push-buttons and hand-switches, are shown as stylized pictures, and are operated via special pop-up menus and dialog boxes in response to user inputs.
This manual assumes that the user is familiar with the main characteristics of water cooled thermal nuclear power plants, as well as understanding the unique features of the BWR.
TABLE I. SUMMARY OF BWR SIMULATOR FEATURES SYSTEM SIMULATION SCOPE DISPLAY
PAGES OPERATOR
CONTROLS MALFUNCTIONS REACTOR
CORE • Neutron flux levels over a range of 0.001 to 110%
full power, 6 delayed neutron groups
• Decay heat (3 groups)
• Reactivity feedback effects - void, xenon, fuel temperature, moderator temperature
• 2 phase flow & heat transfer
• Reactivity control rods
• Essential control loops - Reactor Pressure Control;
Core Recirculation Flow Control; Reactor Power Regulation; Reactor Water Level Control;
Turbine Load/Frequency Control
• Plant Overview
• BWR
Reactivity &
Setpoints
• BWR Power /Flow Map &
Controls
• Reactor power and rate of change (input to control
computer)
• Manual control of control rods
• Reactor scram
• Manual Control Rods “run-in”
• Manual control
of core recirculation
flow rate
• Manual
adjustment of reactor water control level setpoint
• Increasing and decreasing core flow due to Flow Control
malfunctions
• Inadvertent
withdrawal of one bank of control rods
• Inadvertent
insertion of one bank of control rods
• Inadvertent reactor isolation
• Power loss to 3 Reactor Internal Pumps (RIPs)
• Reactor bottom break
STEAM &
FEED- WATER
• Steam supply to turbine and reheater
• Main Steam Isolation Valve
• Turbine Bypass to condenser
• Steam Relief Valves to Suppression Pool in containment
• Extraction steam to feed heating
• Feedwater system
• BWR Feedwater and Extraction Steam
• Reactor water level setpoint changes:
computer or manual
• Extraction
steam to feedwater
heating
isolating valves controls
• Deaerator main steam extraction pressure control
• Feed pump
on/off controls
• Loss of both feedwater pumps
• Loss of feedwater heating
• Reactor
feedwater level control valve fails open
• Safety valves on one main steam line fail open
• Steam line break inside Drywell
• Feedwater line break inside Drywell
SYSTEM SIMULATION SCOPE DISPLAY
PAGES OPERATOR
CONTROLS MALFUNCTIONS TURBINE-
GENERAT OR
• Simple turbine model
• Mechanical power and generator output are proportional to steam flow
• Speeder gear and governor valve allow synchronized and non- synchronized operation
• BWR Turbine- Generator
• Turbine trip
• Turbine run- back
• Turbine run-up and
synchronization
• Turbine
Speeder Gear control: manual or computer control
• Steam Bypass Valve
Computer or Manual Control
• Turbine throttle pressure
transmitter fails low
• Turbine trip with Bypass Valve failed closed
• Increasing and decreasing steam flow due to Pressure Control System failures
OVERALL
UNIT • Fully dynamic interaction between all simulated systems
• Turbine-Following-
Reactor load maneuvering
• Unit annunciation
• Major control loops
• BWR Plant Overview
• BWR
Reactivity &
Setpoints
2.1 SIMULATOR STARTUP
• Select program ‘BWR’ for execution - the executable file is BWR.exe
• Click anywhere on ‘BWR simulator” screen
• Click ‘OK’ to ‘Load Full Power IC?’
• The simulator will display the ‘Plant Overview’ screen with all parameters initialized to 100% Full Power
• At the bottom right hand corner click on ‘Run’ to start the simulator
2.2 SIMULATOR INITIALIZATION
If at any time you need to return the simulator to one of the stored initialization points, do the following:
• ‘Freeze’ the simulator
• Click on ‘IC’
• Click on ‘Load IC’
• Click on ‘FP_100.IC’ for 100% full power initial state
• Click ‘OK’ to ‘Load C:\BWR_Simulator\FP_100.IC’
• Click ‘YES’ to 'Load C:\BWR_Simulator\FP_100.IC’
• Click ‘Return’
• Start the simulator operating by selecting ‘Run’.
2.3 LIST OF BWR SIMULATOR DISPLAY SCREENS (1) BWR Plant Overview
(2) BWR Control Loops
(3) BWR Power/Flow Map & Controls (4) BWR Reactivity & Setpoints
(5) BWR Scram Parameters (6) BWR Turbine Generator
(7) BWR Feedwater & Extraction Steam (8) BWR Trends
2.4 GENERIC BWR SIMULATOR DISPLAY COMMON FEATURES
The generic BWR simulator has 8 interactive display screens or pages. Each screen has the same information at the top and bottom, as follows:
• The top of the screen contains 21 plant alarms and annunciations; these indicate important status changes in plant parameters that require operator actions;
• The top right hand corner shows the simulator status:
⇒ The window under ‘Labview’ (this is the proprietary graphical user interface software that is used to generate the screen displays) has a counter that is incrementing when Labview is running; if Labview is frozen (i.e. the displays cannot be changed) the counter will not be incrementing;
⇒ The window displaying ‘CASSIM’ (this is the proprietary simulation engine software that computes the simulation model responses) will be green and the counter under it will not be incrementing when the simulator is frozen (i.e. the model programs are not executing), and will turn red and the counter will increment when the simulator is running;
• To stop (freeze) Labview click once on the ‘STOP’ (red “Stop” sign) at the top left hand corner; to restart ‘Labview’ click on the ⇒ symbol at the top left hand corner;
• To start the simulation click on ‘Run’ at the bottom right hand corner; to ‘Stop’ the simulation click on ‘Freeze’ at the bottom right hand corner;
• The bottom of the screen shows the values of the following major plant parameters:
⇒ Reactor neutron power (%)
⇒ Reactor thermal power (%)
⇒ Turbine generator output power (Gross) (%)
⇒ Reactor pressure (KPa)
⇒ Core flow (Kg/s)
⇒ Reactor water level (m)
⇒ Balance of plant (BOP) steam flow (Kg/s) — that means steam flow after the main steam isolation valve
⇒ Feedwater flow (Kg/s)
⇒ Average fuel temperature (Deg. C)
• The bottom left hand corner allows the initiation of two major plant events:
⇒ ‘Reactor trip’ or ‘reactor scram’
⇒ ‘Turbine trip’
These correspond to hardwired push buttons in the actual control room.
• The box above the trip buttons shows the display currently selected (i.e. ‘plant overview’); by clicking and holding on the arrow in this box the titles of the other displays will be shown, and a new one can be selected by highlighting it;
• The remaining buttons in the bottom right hand corner allow control of the simulation one iteration at a time (‘Iterate’); the selection of initialization points (‘IC’); insertion of malfunctions (‘Malf’); and calling up the ‘Help’ screen (online hyperlinked “Help” is not available yet).
As a general rule, all dynamic display values shown in display boxes on the screens follow the following conventions:
• All pressure values are designated as “P” next to the display box, and have units of KPa;
• All temperature values are designated as “T” next to the display box and have values of deg. C;
• All flow values are designated as “F” next to the display box and have values of Kg/s;
• 2 phase qualities are indicated as “X” next to the display box and have% as units.
Valve status and pump status as shown by dynamic equipment symbols are represented as follows:
• Valve status — red for valve fully open; green for valve fully closed; partial red and green indicates partial valve opening;
• Pump status — red for running; green for stopped.
3. SIMULATOR DISPLAY SCREENS 3.1 BWR PLANT OVERVIEW SCREEN
This screen shows a ‘line diagram’ of the main plant systems and parameters. No inputs are associated with this display. The systems and parameters displayed are as follows (starting at the bottom left hand corner):
• REACTOR is a point kinetic model with six groups of delayed neutrons; the decay heat model uses a three-group approximation; 2-phase flow and heat transfer. Reactivity calculations include reactivity control rods — FMCRD, fine motion control rods, and reactivity feedback effects due to Xenon, 2-phase voiding in channels; fuel temperature (Doppler) and moderator (light water) temperature.
• The reactor parameters displayed are:
Reactor dome section
⇒ Dome steam temperature (ºC)
⇒ Dome pressure (p)
⇒ Steam flow from core (Kg/s)
⇒ Reactor water level (m) Reactor core section
⇒ Neutron power rate (%/ second)
⇒ Thermal power generated by core (MW(th))
⇒ Average fuel temperature (ºC)
⇒ Coolant flow rate in core (Kg/s)
⇒ Coolant pressure at core exit (p)
⇒ Coolant temperature at core exit (ºC)
⇒ Coolant quality at core exit (X%)
⇒ Control rods position in core (% of total length in core). Note control rods reactivity worth is as follows: 100% in core - negative 100 milli-K; 100% out-of-core - positive 70 milli-K.
Reactor downcomer section
⇒ Reactor internal pumps head (KPa)
⇒ Reactor internal pumps speed (RPM)
• Outside the reactor pressure vessel (RPV) and still inside the containment are shown:
⇒ Main steam isolation valve status: red means fully open;
⇒ The main steam lines have branch connection to the safety relief valves (SRVs) that are connected to the suppression pool inside containment. Here all the SRVs are shown in “one equivalent valve” symbol; in fact there are 8 SRVs, with 2 SRVs associated with each main steam line; and there are four separate main steam lines.
So the steam flow shown is for total steam flow through all the SRVs.
⇒ Emergency core cooling (ECC) injection is shown here as “core spray” flow in case of loss of coolant accident. Note in this model, no distinction is made between LP flooder and HP flooder. They are all treated as one source and go directly to the core upper plenum. The ECC water is assumed to be at temperature ~ 20 deg. C.
⇒ Note: The containment drywell and wetwell are not modeled in this simulator. But in the event of major accidents inside the drywell, such as feedwater line break, steam line break, and reactor bottom break, these breaks will cause high pressure in drywell, which in turn will trigger the LOCA signal. As a result, ECC will be started, the reactor will be scrammed, and “isolated”.
• Outside containment is the balance of plant systems — turbine generator, feedwater &
extraction steam. The following parameters are shown:
⇒ Status of control valves is indicated by their colour: green is closed, red is open; the following valves are shown for the steam system:
Turbine governor valve opening (%) Steam bypass valve opening (%)
⇒ Moisture separator and reheater (MSR) drains flow (kg/sec)
Generator output (MW) is calculated from the steam flow to the turbine
• Condenser and condensate extraction pump (CEP) are not simulated but the pump status is shown.
• Simulation of the feedwater system is very much simplified; the parameters displayed on the plant overview screen are:
⇒ Total feedwater flow to the steam generators (kg/sec)
⇒ Average feedwater temperature after the high pressure heaters (HPHX)
⇒ Status of feedwater pumps (FWP) is indicated as red if any pumps are ‘ON’ or green if all the pumps are ‘OFF’
Three trend displays show the following parameters:
• Reactor neutron power, reactor thermal power and turbine power (0-100%)
• Core flow, steam flow, feedwater flow (Kg/s)
• Reactor pressure (kPa)
The upper and lower limits of the parameter trends can be altered while the simulator is running by clicking on the number to be changed and editing it.
Note that while the simulator is in the ‘Run’ mode, all parameters are being continually computed and all the displays are available for viewing and inputting changes.
3.2 BWR CONTROL LOOPS SCREEN
This screen shows all the essential control loops for the generic BWR plant, and the essential control parameters for these loops. The parameters are:
• Generator output and frequency
• Feedwater flow
• Reactor pressure
• Reactor water level
• Neutron flux
• Core flow
The essential control loops are:
⇒ Control rods control — press the button to display a pop-up window, which describes the functions of the control system. The control rod drive system is composed of three major elements: the fine motion control rod drive, FMCRD mechanisms; the hydraulic control unit (HCU) assemblies; the control rod drive hydraulic subsystem (CRDH). The FMCRDs together with the other components are designed to provide:
(1) Electric-motor-driven positioning for normal insertion and withdrawal of the control rods;
(2) Hydraulic-powered rapid control rod insertion (scram) in response to manual or automatic signals from the reactor protection system (RPS);
(3) Electric-motor-driven "Run-Ins" of some or all of the control rods as a path to rod insertion for reducing the reactor power by a sizable amount.
For manual control of control rods and recirculation pumps, go to Screen BWR Power/Flow Map & Controls".
⇒ Reactor power control — press the button to display a pop-up window, which describes the functions of the control system. The reactor power output control system consists of control rods, rod drive system and recirculation flow control system. The control rods and their drive system maintain a constant desired power level by adjusting the position of the rods inside the core. The recirculation flow control also controls the reactor power level by changing the recirculation flow to alter the void density of the two-phase water in the core, which leads to a change in reactivity in the core due to the altered neutron moderation efficiency of the coolant. The recirculation flow is controlled by recirculation pumps known as reactor internal pumps (RIPs). The pump speed changes according to the change of frequency of the induction motor that drives the pump. Different pump speed will give rise to different pump dynamic head in the core recirculation flow path, resulting in different core flow. This recirculation flow control system is capable of changing the reactor output rapidly over a wide range. Go to Screen
"BWR Reactivity & Setpoints" for changing reactor power setpoint, and observe the Power & Recirculation Flow relationship in Screen " BWR Power/Flow Map &
Controls"
⇒ Reactor pressure control — press the button to display a pop-up window, which describe the functions of the control system. When the reactor is in power level operation, the reactor pressure is automatically controlled to be constant. For that purpose, a pressure controller is provided and is used to regulate the turbine inlet steam pressure by opening and closing the turbine governor control valve and the turbine bypass valve. Currently, the reactor pressure setpoint is set at plant design pressure of 7170 KPa.
⇒ Reactor water level control — press the button to display a pop-up window, which describes the functions of the control system. In order to suppress the water carry-over in the steam going to the turbine as well as to prevent the core from being exposed, three signals detecting the feedwater flow, the main steam flow, and the water level inside the reactor pressure vessel are provided. The flow of feedwater is automatically controlled to maintain the specified water level by a "three element" control scheme: steam flow, feedwater flow, and water level. The valve opening of the feedwater control valve provided at the outlet of the feedwater pumps is regulated by the control signal as result of this "three-element" control scheme. To modify the reactor water level setpoint, go to Screen " BWR Feedwater & Extraction Steam", and call up the related Pop-Up Window.
⇒ Turbine control — the turbine control employs an electrohydraulic control system (EHC) to control the turbine valves. Under normal operation, the reactor pressure control (RPC) unit keeps the inlet pressure of the turbine constant, by adjusting the opening of the turbine “speeder gear” which controls the opening of the turbine governor valve opening. Should the generator speed increase due to sudden load rejection of the generator, the speed control unit of the EHC has a priority to close the turbine governor valve over the reactor pressure control (RPC) unit.
⇒ Turbine steam bypass system — the simulated BWR plant is designed with turbine steam bypass capacity of over 75% rated steam flow. Hence, in the event of any reactor pressure disturbances, either caused by reactor power sudden increases, or due to turbine load rejection, or frequency changes, and the reactor pressure control unit cannot cope with these pressure increases fast enough, the turbine bypass valve will open up to pass steam to condenser to reduce sudden reactor pressure increases. The setpoint for the bypass valve to “come in” when the turbine is not “tripped” is — 130 KPa (called bias) over the normal reactor setpoint of 7170 KPa. That means the bypass valve will not open until the reactor pressure increases to > 7300 KPa; this gives room for the turbine control valve to act in an attempt to control pressure back to 7170 KPa. However, if the turbine is tripped, the bias will be removed and the setpoint for the bypass valve is 7170 KPa. Therefore, with the designed Steam Bypass System, reactor scram is not necessary for spurious turbine trip, but a large reduction of reactor power from full power to ~ 60% FP will result due to “Rods Run-in” coming into action. This term refers to rapid insertion of control rods into the core to reduce reactor power by a large step. However, for full load rejection, a reactor scram will occur.
3.3 BWR POWER/FLOW MAP & CONTROLS This screen shows
(a) The relationship between reactor neutron power versus core flow;
(b) The reactor core conditions with respect to boiling height; water level; fuel temperature;
coolant temperature, pressure and flow; steam pressure, flow and temperature;
(c) Controls for scramming the reactor, as well as for resetting the scram; the AUTO/MANUAL controls for the control rods (FMCRD) and for the reactor internal pumps (RIPs) drive unit.
⇒ POWER/FLOW MAP
• The power flow map is a representation of reactor power vs. recirculation flow. The horizontal axis is the core flow in% of full power flow. The vertical axis is reactor neutron power in% full power.
• Any operation path that changes the power and the flow from one condition to another condition through control rod maneuver and/or recirculation flow change can be traced on this map. Under normal plant start-up, load maneuvering, and shutdown, the operation path through REGION IV is recommended. In fact, the line which borders between REGION I & IV, REGION III & IV, the blue region and REGION IV is the
“maximum power-flow” path to be followed for power increases and decreases, and usually operation of the plant is “below” this “maximum” power-flow line.
• Limits are imposed to prevent operation in certain areas of the Power - Flow Map - (1) To maintain core thermal limits and to avoid operation above licensed power level
- there are three measures to prevent that:
(a) Control rods withdrawal “Blocked” — if at any time, the current power exceeds 105% of the power designed for the current flow rate (in accordance with the maximum power-flow line as described above), the Control Rods withdrawal will be “blocked” until the power drops to 5% less than the current value. Should this occur, the alarm “Hi Neut Pwr vs Flow” will be in “Yellow” color, as well, in the BWR Reactivity & Control Screen, there will be a “yellow color message” saying “Controls Rods Out Blocked”.
(b) Control rods “Run-in” — if any time the current power exceeds 110% of the power designed for the current flow rate (in accordance with the maximum power-flow line as described above), the control rods will be inserted into the core to reduce power quickly and the “Rods Run-in” will be stopped until the power has been reduced to 10% less than the current value.
Should this occur, the alarm “Hi Neut Pwr vs Flow” will be in “Yellow”
color, as well as the alarm “Rods Run-in Req’d”.
(c) Reactor scram — if any time the current power exceeds 113% of the power designed for the current flow rate (in accordance with the maximum power- flow line as described above), the reactor will be scrammed.
(2) To avoid operation where core instability may occur — in REGION III.
It is a well-known and well-documented phenomenon in the BWR that oscillations in neutronic and thermal-hydraulic parameters occur during operation in the conditions “low flow - high power” region identifiable in the Power/Flow Map as REGION III.1
Research has shown that such oscillations are characterized by “density wave”
oscillations. From a physical point of view, the removal of thermal power by boiling water in a vertical channel, in a closed or open loop configuration, may cause instability in the operation owing to density changes and various thermalhydraulic feedback mechanisms. Since the coolant is also a neutron moderator, an oscillation in the coolant density (void content) is reflected as a variation of the thermal neutron flux, which in turn, via the heat flux, affects the void. This may cause a coupled neutronic-thermalhydraulic oscillation under certain power and core flow conditions. The details of core instability in Region III belong to an advanced topic that is beyond the scope of this manual.
(3) To avoid operation where excess moisture in the steam may be carried to the main turbine — in REGION II.
⇒ REACTOR CORE GRAPHICS
The right side of the screen depicts the reactor core conditions at all operations. As well, the control devices for control rods and the reactor internal pumps are provided. Starting from the bottom:
⇒ FMCRD auto/manual button — this button when pressed will allow the user to switch the control rods to be under the “automatic” control scheme or under “manual”
control. If they are in “manual” mode, the switch status will be indicated as “MAN”, and the user can then control the rods by pressing the button above the designated number of the control rods bank #1 to #8 respectively. A control pop-up will appear when the button is pressed, allowing the user to “insert” or to “withdraw” each “bank” of rods separately, by using the “in”, or “out” pushbutton respectively in the pop-up. To stop the movement of the rods, use the “stop” pushbutton in the pop-up.
When the FMCRDs are in “Auto”, the automatic control scheme is in control, and its details are described in the “BWR Reactivity & Controls Screen” section. In Auto mode, all the controls rods move together as controlled by the reactor power regulating system.
Note:
• There are approximately 208 FMCRDs in total, they are positioned and calibrated with reactivity worth of -100 mk when all of them are 100% in core, and +70 mk when all of them are 100% out of core; 0 mk when they are ~ 41% in core.
1 OECD Report OCDE/GD(97)13: “State of the Art Report on boiling water reactors stability (SOAR on BWRS), January 1997.
• For the purpose of this generic simulator, the rods are grouped in 8 banks, so each bank of rods has + 8.75 mk when fully out of core; and -12.5 mk when fully in core. The full speed travel time for the rod movement during power maneuvering is typically 60 sec., or considering for the total FMCRDs in Auto mode, where all the rods move together, the reactivity change rate is ~ 2.8 mk per sec. Considering moving the banks of rods individually under manual Mode, then the reactivity change rate for each bank under manual mode control is ~ 0.36 mk per sec.
• The FMCRDs will be fully inserted into the core in the event of a reactor scram.
In such case, the fast insertion time is typically 3 sec. for 100% insertion.
• The average rods position in core is shown on the right hand bottom corner.
⇒ SCRAM status indication, manual scram/reset button, SCRAM reset button — when the reactor is scrammed, and if scram conditions still exist, there will be a “YES”
sign next to the “SCRAM ST” indicator. As well, the alarm “Reactor Scram” is in
“Red” color. Assuming the scram conditions have already disappeared, and user wishes to reset the scram, the button to the right of the “YES” indicator is pressed, which will bring up a control pop-up. The user can then press the “reset” pushbutton on the pop-up.
If the reactor scram conditions do not exist at that time, then the “YES” sign will be changed to “NO” sign, meaning that the SCRAM Status indicates “NO” scram conditions. At this point, the user can proceed to press the “SCRAM RESET” button on the left side of the “SCRAM ST” indicator. When this button is pressed, the “Reactor Scram” Alarm will disappear, and the rods withdrawal will begin, as can be seen from the downward arrows shown for the rods banks. The rods withdrawal will stop at the
“reset line” which is 50% in core, pending on control actions taken by the reactor regulating system. At the 50% in core position, the rods reactivity is ~ -15 mk, so the reactor is still subcritical. It is the role of the reactor regulating system to bring the reactor back to criticality.
⇒ ON/OFF control for RIP pump motor — there are 10 reactor internal pumps (RIPS), but they are modeled in one “lumped” pump, so the ON/OFF control button is used to turn “ON” or “OFF” all the pump motors. When the motor power is “Off”, the speed drive will go to the minimum position, giving “zero” pump head. When the power is
“ON”, the speed drive signal is subject to the flow controller signal that is described later. The RIPs speed changes according to the change of frequency of the induction motor that drives the pumps. Different pump speed will give rise to different pump dynamic head in the core recirculation flow path, resulting in different core flow. The automatic flow control scheme is handled by a flow controller. First, based on the reactor power setpoint, there is a pre-programmed flow rate schedule according to that power setpoint. The pre-programmed schedule is typically to follow the “maximum power — flow” path as described in POWER-FLOW map section. Given the flow setpoint, the flow controller will drive the speed drive to provide enough pump head until the desired flow rate is achieved.
⇒ Average pump head indicator, average pump speed indicator, flow controller control button — the average pump head is shown in KPa, and the average pump speed is shown in RPM. The flow controller button is labeled as RIPCrl. When this button is pressed, it will show a typical controller template. The setpoint is on remote setpoint (RSP), meaning that it receives setpoint from the pre-programmed flow rate scheduler.
The horizontal “blue” bar is for indication of the current flow rate; the “green” pointer is flow setpoint indicator. When the flow rate is at setpoint, the “green” pointer can be
seen to be at the tip of the “blue” bar. When the “auto” button is pressed, the controller is no longer subject to remote setpoint, rather it will be subject to “local” setpoint, and the user can enter a “new” setpoint into the box under “SP”. Alternatively, the user can switch the controller to manual. In that case, the output% box will show
“increase/decrease” arrow, meaning a% value can be entered into the out(%) box to change controller output directly, which will go to the speed drive.
⇒ Core conditions display - the following parameters are shown for the core conditions:
• Average fuel temperature
• Coolant flow rate, temperature at core exit, and quality (%) at core exit; feedwater flow rate; coolant recirculation ratio “R” are shown. The “blue” arrows show the flow path of the coolant out from the core channels, as it goes to the core upper plenum, enters the dome space, mixes with incoming feedwater, and goes down to the downcomer, to enter the suction path of the reactor internal pumps. Then from the RIPs discharge, the coolant enters the core lower plenum generally with subcooled temperature. As the coolant enters the core channel again, the subcooled water receives heat from the fuel bundles, and becomes two-phase fluid that exits the core with quality. The steam vapor from the coolant escapes the fluid and becomes saturated steam; the remaining water content of the fluid is recirculated back to the downcomer after mixing with incoming feedwater.
• The boiling height – 2-phase boiling region of the core is shown in a “pink” color.
It is animated, so as the boiling height changes as the core conditions change, the
“pink” section boundary changes. The same applies to the “light blue” subcooled section - or non-boiling section of the core.
• The water level in the core is indicated as a” blue” color and is animated. As the level changes, the “blue” section boundary changes.
• For the Dome space, the gray arrows show the flow path of saturated steam, where the flow, pressure and temperature are shown.
3.4 BWR REACTIVITY & CONTROLS SCREEN
This screen shows input devices which facilitate reactor power setpoint entry, as well as to facilitate reactor ‘Manual Scram’, or ‘Manual Rods Run-in”. These inputs interact with an underlying reactor regulating system.
• The user can enter new reactor power target and power change rate by pressing the button located near the bottom left side of the screen next to “RCTR PWR SETPOINT”.
When this button is pressed, a control pop-up will allow the user to enter the reactor power target in%, and the rate in% full power per sec. (if current power is > 20% FP), or% present power per sec. (if the current power is < 20% FP). The purpose is to allow higher power rate change only at higher power.
• After the setpoint and rate are entered, the “ACTUAL SETPOINT” section reflects the setpoint actually accepted by the regulating system. Then the incremental demanded setpoint is computed in the “DEMANDED SETPOINT” section; as well the rate is shown in “DEMANDED RATE SETPOINT” section. The POWER ERROR is computed as:
POWER ERROR = ACTUAL POWER - DEMANDED POWER
• The reactor regulating system will check if the current power is < 65%. If it is, then the control rods movement is necessary. Based on the power error - whether it is positive or negative, the rods will be inserted or withdrawn accordingly so that the power error becomes zero. If the current power is > 65%, then usually rods movement is not required, the new incremental demanded power setpoint signal in turn is sent to the flow rate scheduler (as described in previous section) which will provide a flow rate setpoint to the flow controller. If the flow rate increase/decrease cannot provide enough reactivity change causing sufficient reactor power increase/decrease so that the power error is less than a pre-determined dead-band, the rods movement will become necessary at that time so that the power error is within limits.
• The display information on this screen provides the important information regarding reactivity changes as shown by the various reactivity feedback effects - void density;
xenon; fuel temperature, coolant temperature, as well as the control rods reactivity changes as a result of their movement in the core. Note that reactivity is a computed not a measured parameter, it can be displayed on a simulator but is not directly available at an actual plant. Also note that when the reactor is critical the total reactivity must be zero.
• Note that the BWR plant is always operating with turbine-following-reactor mode.
• The buttons at the top of the screen allow the user to perform a “manual” “rods run-in”, as well as “manual” reactor scram.
• The “HOLD POWER” button near the top left hand corner allows the user to “suspend”
reactor power changes at any time. Just pressing the button once will result in the Demanded Power Setpoint being set to “frozen”, if it was increasing or decreasing initially.
• Near the bottom of the middle section of the screen is the button that can switch the controls rods “AUTO/MANUAL”.
3.5 BWR SCRAM PARAMETERS SCREEN
This screen shows all the parameters that will cause reactor scrams:
• High neutron flux/low core flow - as described previously, if at any time the current power exceeds 113% of the power designed for the current flow rate (in accordance with the maximum power-flow line as described above), the reactor will be scrammed.
• High drywell pressure/LOCA detected — if the drywell pressure exceeds 103 KPa, then the LOCA logic senses that a LOCA condition has occurred.
• Reactor level low — the scram setpoint is 11 meters above reactor bottom. Normal level is 13.5 meters above reactor bottom.
• Reactor pressure high — the scram setpoint is 7870 KPa. Normal reactor pressure is 7170 KPa.
• Reactor level very high — the scram setpoint is 14.5 meters above reactor bottom.
• Main steam isolation valve closed/reactor isolated.
• Main steam line radioactivity high.
• Turbine power/load unbalance or loss of line (load rejection).
• Earthquake acceleration large.
• Manual scram.
3.6 BWR TURBINE GENERATOR SCREEN
This screen shows the main parameters and controls associated with the turbine and the generator. The parameters displayed are:
• Reactor side main steam pressure and main steam flow (before the isolation valve);
main steam isolation valve status
• Main steam header pressure after the main steam isolation valve.
• Status of main steam safety relief valves (SRVs)
• Status, opening and flow through the steam bypass valves
• Steam flow to the turbine (kg/sec)
• Governor control valve position (% open)
• Generator output (MW)
• Turbine/generator speed of rotation (rpm)
• Generator breaker trip status
• Turbine trip status
• Turbine control status
• All the trend displays have been covered elsewhere or are self explanatory
The following pop-up menus are provided:
• TURBINE RUNBACK — sets target (%) and rate (%/sec) of runback when ‘Accept’ is selected
• TURBINE TRIP STATUS — trip or reset
• Steam bypass valve ‘AUTO/MANUAL’ control — AUTO select allows transfer to MANUAL control, following which the manual position of the valve may be set.
• Computer control of the speeder gear must be in the 'Enable' mode since manual control of the speeder is not provided in this version of the simulator.
• Turbine runup/speedup controls
3.7 BWR FEEDWATER AND EXTRACTION STEAM SCREEN
This screen shows the portion of the feedwater system that includes the deaerator, the boiler feed pumps, the high pressure heaters and associated valves, with the output of the HP heaters going to the reactor water level control valves. The following parameters are displayed:
• Main steam header pressure after the main steam isolation valve, steam flow through the turbine governor valve and the bypass valve.
• Deaerator level (m) and deaerator pressure (KPa); extraction steam motorized valve status and controls from turbine extraction, as well as pressure controller controls for main steam extraction to deaerator. The extraction steam flows are shown respectively for turbine extraction as well as for main steam extraction to the deaerator.
• Main feedwater pump and auxiliary feedwater pump status with associated pop-up menus for ‘ON/OFF’ controls.
• HP heater motorized valves MV2 and MV3 and pop-up menus for open and close controls for controlling extraction steam flow to the HP heaters.
• Flow rate at reactor level control valve outlet and feedwater temperature.
4. SIMULATOR EXERCISES
4.1 INTRODUCTORY EXERCISES
4.1.1 Power maneuver: 10% power reduction and return to full power
• Initialize simulator to 100% full power
• Verify that all parameters are consistent with full power operation.
⇒ Go to “Reactivity & Setpoint” Screen
⇒ Press RCTR PWR SETPOINT button
⇒ In pop-up menu lower ‘target’ to 90.00% at a ‘Rate’ of 1.0%/sec
⇒ ‘Accept’ and ‘Return’
• Observe the response of the displayed parameters until the transients in reactor power and steam pressure are completed. Observe the power flow path on “ Power Flow Map
& Controls” screen.
• Continuing the above operation, raise “UNIT POWER” to 100% at a rate of 1.0%FP/sec.
4.1.2 Reduction to 0% full power and back to 100% full power
• Initialize the simulator to 100%FP, reduce power using 25% steps at 1%/sec and record the following values:
Parameter Unit 100% 75% 50% 25% 0% Comments
Reactor Power %
Core Flow Rate Kg/s
Coolant Temperature
°C Coolant Pressure at core exit
KPa Coolant Quality at core exit
%
Reactor Level m
Reactor Steam Pressure
KPa
Reactor Steam Flow Kg/s
Feedwater Flow kg/s
Turbine-Generator Power
%
Under “Comments” please note type of parameter change as a function of reactor power 0%
→ 100% FP: constant, linear increase or decrease, non-linear increase or decrease. In particular, comment on the power — flow path during the power evolution, and comment if the path enters into particular regions in the power flow map.
• Increase reactor power back to 100% after ~ 0% is reached, using 25% steps and 1% per sec rate. Repeat the above recordings and comments.
4.1.3 Turbine trip and recovery
From a simulator Initial state of 100% full power, press the turbine trip button on the left hand bottom corner of the screen, and confirm turbine trip.
• Notice the power flow path on “Power Flow Map and Controls” screen
• What is reactor power when turbine speed settles at 5 rpm?
• What is the steam flow through the bypass valve on the turbine generator screen?
• What is the reactor pressure?
• Go to BWR turbine generator screen, reset turbine trip, select ‘TRU ENABLE’, and select “TRU Speedup” to synchronize the generator and load to about 10%FP;
• After turbine is in service, what happens to the steam bypass valve as the turbine power increases? Note the reactor pressure.
• After the turbine power is equal to reactor power, go to “Reactivity & Setpoints” to increase reactor power to 100% in 25% steps at 1% per sec.
4.1.4 Reactor scram and recovery
• Initialize the simulator to 100%FP
• Manually scram the reactor
• Observe the response of the overall unit. Describe and explain the following responses upon reactor scram:
(a) Recirculation core flow;
(b) Reactor pressure;
(c) The sub-cooled & boiling region boundary of core — explain the changes;
(d) Turbine load & bypass system.
• Wait until generator power is zero and reactor neutron power ~ 0.1%. Or alternatively instead of waiting, recall the Zero Power COLD IC “Zero_cld.ic” from the IC menu, and re-start from this IC plant condition.
• Reset reactor scram using control devices provided on “Power Flow Map & Controls”
screen. Press the YES button next to SCRAM ST. (Reminder, after pressing the reactor reset pushbutton, it is necessary to press the SCRAM RESET button to begin rods withdrawal).
• Record the time (using the display under the chart recorders) needed to withdraw all rods to Reset line.
• Go to “Reactivity & Setpoint” Screen; record the net total reactor reactivity.
• Is the reactor subcritical, critical or supercritical?
• Enter a new power setpoint say 10%, at a rate of 10% present power per sec. and observe any changes in control rods reactivity. The rods should be withdrawing.
Otherwise, go to the power flow map & controls screen, and press the “SCRAM RESET” Button again, you should see the “down” arrows showing the rods movement.
• Observe the response of the reactor regulating system and the reactivity changes that take place
• Continue to raise power to 60% FP in 25% steps at 10% present power per sec. Note after 20% FP is reached, you can enter 1% FP per sec. as the rate.
• If the turbine is tripped as a result of low power, reset turbine trip, synchronize and reload as described in previous section.
4.2 MALFUNCTION EXERCISES
4.2.1 Loss of feedwater - both FW pumps trip.
Go to BWR Feedwater & Extraction Steam Screen. Load the 100% FP IC, then insert the above malfunction. This malfunction leads to total loss of feedwater to the Reactor Pressure Vessel.
When this malfunction transient occurs:
• On BWR Feedwater & Extraction Steam Screen, observe that both the feedwater pumps stop.
• Go to Power/Flow Map Screen. The reactor level drops quickly due to loss of feedwater flow.
• Dome pressure is decreasing gradually as reactor water level drops. Provide explanation why this is happening.
• The drop in dome pressure will cause more boiling in core (why?), hence the steam quality increases; same with void fraction. Increase in void fraction will result in more negative reactivity. Thus reactor power decreases.
• As dome pressure drops, the turbine inlet pressure also drops. In order to restore the Reactor Pressure at setpoint, the Reactor Pressure Controller closes the turbine governor valve slightly. As a result, the generator MW drops.
• As reactor power decreases, the Reactor Power Controller will try to increase power by increasing recirculation flow, as well as by withdrawing the control rods out of core
• Note the movement of the yellow cursor in the Power/Flow Map.
• After a short while, the Reactor will be scrammed by low water level (11 M above the reactor bottom).
• Emergency Core Cooling (ECC) will be activated on a low reactor level, and the reactor will be isolated.
4.2.2 Increasing core flow due to flow control failure
Go to Power/Flow Map. Load the 100% FP IC. Run the simulator. Note that this malfunction causes the recirculation flow controller to fail in such a way that the process variable input for the controller fails in “low” value. But the flow transmitter reading for display is normal. The consequence is that the Recirculation Flow Control System is “fooled” into thinking that the recirculation flow is lower than the flow setpoint, hence it will increase the speed of the Reactor Internal Pumps (RIPs) to increase flow.
Now insert the malfunction. When this malfunction transient occurs:
• On the Power/Flow Map Screen, observe the movement of the yellow cursor.
• Observe that the coolant flow is increasing. As coolant flow increases, core quality X (%) decreases. Provide explanation.
• As core quality decreases, so does void fraction. As a result, the reactor neutron flux increases. Provide explanation.
• As soon as the reactor power exceeds the target setpoint, the Reactor Power Control system will attempt to decrease reactor power by inserting control rods.
• Observe that the cursor moves beyond the operating path on the Power/Flow Map.
4.2.3 Decreasing core flow due to flow control failure
Go to Power/Flow Map Screen. Load the 100% FP IC. Run the simulator. Note that this malfunction causes the recirculation flow controller to fail in such a way that the process variable input for the controller fails in “high” value. But the flow transmitter reading for display is normal. The consequence is that the Recirculation Flow Control System is “fooled”
into thinking that the recirculation flow is higher than the flow setpoint, hence it will decrease the speed of the Reactor Internal Pumps (RIPs) to decrease flow.
Now insert the malfunction. When this malfunction transient occurs:
• On Power/Flow Map screen, observe the movement of the yellow cursor.
• Observe that the coolant flow is decreasing. As coolant flow decreases, core quality X (%) increases. Provide explanation.
• As core quality increases, so does void fraction. As a result, the reactor neutron flux decreases. Provide explanation.
• As soon as the reactor power is lower than the target setpoint, the Reactor Power Control system will attempt to increase reactor power by withdrawing control rods.
• But the rate of flow decrease is faster than the reactor power adjustment; hence the cursor is above the normal operating path on the Power/Flow Map. Observe the alarm
“Hi Neut Pwr vs Flow”.
• As the malfunction event evolves, the cursor moves into the “unstable region III” and sits on top of it. Observe that the rods are almost fully withdrawn.
• The reactor pressure is decreasing and low-pressure alarms come on. Explain why Reactor Pressure Control cannot increase reactor pressure back to setpoint.
• As well, there is an alarm “ Turbine runback”. Explain why there is a turbine runback.
4.2.4 Decreasing steam flow from dome due to pressure control failure
Go to the Power/Flow Map. Load the 100% FP IC. Run the simulator. Note that this malfunction causes the Reactor Pressure Controller to fail in such a way that the process variable input for the controller fails in “low” value. But the pressure transmitter reading for display is normal. The consequence is that the Reactor Pressure Control System is “fooled”
into thinking that the reactor pressure is lower than the pressure setpoint of 7170 KPa, therefore it will increase coolant flow rate to raise power in order to increase steam flow. At the same time, the Turbine Control system, thinking that main steam pressure is “low”, will decrease opening of the turbine governor valve. Hence steam flow to turbine will decrease.
Thus the generator load is decreasing.
Now insert the malfunction. When this malfunction transient occurs:
• On the Power/Flow Map Screen, observe the movement of the yellow cursor and record the reactor pressure reading - increasing or decreasing, as the malfunction event evolves.
• Observe that the coolant flow is increasing. Provide explanation.
• As coolant flow increases, core quality X (%) decreases. Provide explanation.
• As core quality decreases, so does void fraction. As a result, the reactor neutron flux increases. Provide explanation.
• As soon as the reactor power is higher than the target setpoint, the Reactor Power Control system will attempt to decrease reactor power by inserting control rods.
• Note the Generator Output (%) reading at the bottom of the screen - increasing or decreasing? Provide explanation. Compare this with the Reactor Thermal power (%).
What is the difference? Where does this power difference go to?
• As the malfunction event evolves, the reactor pressure increases, and an alarm “Reactor Press Hi” will come on. But after a short while, it will disappear, then sometime later it will reappear again. Provide explanation why this is the case.
• As this malfunction further evolves, alarm “Rods Run-in Required” will come on.
Observe the movement path of the yellow cursor. Provide explanation.
• Discuss what should the operator do in this event.
4.2.5 Increasing steam flow from dome due to pressure control failure
Go to Power/Flow Map Screen. Load the 100% FP IC. Run the simulator. Note that this malfunction causes the Reactor Pressure Controller to fail in such a way that the process variable input for the controller fails in “high” value. But the pressure transmitter reading for display is normal. The consequence is that the Reactor Pressure Control System is “fooled”
into thinking that the reactor pressure is higher than the pressure setpoint of 7170 KPa, therefore it will decrease coolant flow rate to decrease power in order to decrease steam flow.
At the same time, the Turbine Control system, thinking that main steam pressure is “high”, will increase opening of the turbine governor valve. But the turbine governor valve is almost at the 100% fully open position.
